State-of-the-Art Reactor Consequence Analyses (SOARCA) Project Accident Analysis

1. State-of-the-Art Reactor Consequence Analyses (SOARCA) ProjectAccident Analysis Presented at the USNRC 20th Regulatory Information Conference Washington, DC March 11, 2008 Randall Gauntt, Sandia National Laboratories Charles Tinkler, USNRC Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company,for the United States Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000.

2. SOARCA Objectives • Perform a state-of-the-art, realistic evaluation of severe accident progression, radiological releases and offsite consequences for important accident sequences • Phenomenologically based, consistent, integral analyses of radiological source terms • Provide a more realistic assessment of potential offsite consequences to replace previous consequence analyses • 1982 Siting Study

3. SOARCA Accident Progression Modeling Approach • Full power operation • Plant-specific sequences with a CDF>10-6 (CDF>10-7 for bypass events) • External events included • Consideration of all mitigative measures, including B.5.b • Sensitivity analyses to assess the effectiveness of different safety measures • State-of-the-art accident progression modeling based on 25 years of research to provide a best-estimate for accident progression, containment performance, time of release and fission product behavior

4. 1982 Siting Study • Evaluated potential consequences relevant to generic siting criteria • Used hypothesized, generalized, source term categories • Based on limited knowledge and bounding rationale • Uncoupled from specific plant design or specific sequences • Consequences dominated by • Source term magnitude and timing • Population density • Emergency response

5. Radiological Source Terms • 1982 Siting Study results were dominated by the SST1 source term • Loss of safety features • Large FP release from core • Severe early reactor and containment failure or bypass • 1982 SST1 characterization (magnitude, timing and frequency) reflected then state of understanding and modeling • Early containment failure modes contemporaneously cited included alpha mode (steam explosion) failure, direct containment heating, hydrogen combustion • Research and plant improvements over 25 years have dramatically altered our view of the early failure modes

6. Severe Accident Improvements • Research/plant improvements provided bases to conclude that some presumed early containment failure modes have been shown to be • negligible/highly improbable • In-vessel steam explosion and alpha mode failure • SERG, Sizewell PRA, Experiments (FARO, KROTOS, TROI) • direct containment heating due to high pressure melt ejection • DCH Issue Resolution, experiments at SNL, ANL, Purdue • or can be prevented by accident management • BWR Mark I liner melt through • Hydrogen control systems • For large dry concrete containments, increased containment leakage is failure mode (vs catastrophic failure of the containment)

7. Preliminary SOARCA Findings • No sequences could be identified which resemble the characteristics of the dominant sequence from the 1982 study sequences • Sequences which were identified have lower frequencies than that assigned to SST1 in 1982 study • All sequences identified could be prevented or significantly mitigated by existing or recently developed plant improvements • Important to realistically treat plant features/capabilities and include in probabilistic assessments • Confirmed by MELCOR analyses and served as the basis for evaluating plant/operator response including the TSC

8. Preliminary SOARCA Findings • Containment failure or bypass sequences are still identified in some plant specific PRA but even in those instances severity of conditions are significantly reduced • Reactor vessel lower head failure delayed even for the most severe (and most remote) of sequences (~ 7- 8 hrs) and much delayed for more likely severe sequences ( ~20+ hrs) • Bypass events are delayed beyond timing of SST1, bypass events also reflect scrubbed releases due to submergence of break (consistent, mechanistic modeling) or fission product deposition in the system piping • These conditions while identified as important in current/past PRA, may now be considered to be more amenable to mitigation because of timing (revealed by integral analyses) and plant capabilities

9. Preliminary SOARCA Findings • Without those mitigation strategies, sensitivity studies indicate a radiological release fraction which is significantly smaller than earlier studies. • Unmitigated sensitivities also result in a delayed release

10. Peach Bottom Atomic Power Station Emergency (B.5.b) Equipment • Portable power source for SRVs and level indication • Manual operation of RCIC without dc power • Portable diesel driven pump (250 psi, 500 gpm) to makeup to RCS, drywell, CST, Hotwell, etc. and provide external spray • Portable air supply to operate containment vent valves • Off-site pumper truck can be used in place of portable diesel driven pump

11. Peach Bottom Atomic Power StationLong-term Station Blackout Without Mitigation Without B.5.b mitigation • Accident progression • Core uncovery in 9 hrs • Core damage in 10 hrs • RPV and containment failure in 20 hrs, start of radioactive release, (liner melt-through or containment head flange leakage) • Time between start of evacuation and radioactive release: ~17 hrs • Offsite radioactive release is relatively small • 1 – 4 % release of volatiles, except noble gases • Release is much less severe than 1982 Siting Study • Accident progression timing and emergency evacuation significantly reduce potential consequences

12. Peach Bottom Atomic Power StationLong-term Station Blackout With Mitigation Swollen Vessel Water Level Response

13. Preliminary Findings Summary • B.5.b measures have potential to prevent or significantly delay core damage • Without B.5.b mitigative measures • Releases are significantly lower than 1982 study • Releases can be significantly delayed • Accident progression timing (long time to core damage and containment failure) and mitigative measures significantly reduce the potential for core damage and/or containment failure

14. Peach Bottom Atomic Power StationLong-term Station Blackout Without Mitigation Swollen Vessel Water Level Response

15. Peach Bottom Atomic Power StationLong-term Station Blackout Without Mitigation

16. Peach Bottom Atomic Power StationLong-term Station Blackout Without Mitigation

17. Surry Nuclear Station Emergency (B.5.b) Equipment/Procedures • 2 diesel-driven high-pressure skid-mounted pumps for injecting into the RCS • 1 diesel-driven low-pressure skid-mounted pump for injecting into steam generators or containment • Portable power supply for restoring indication • Portable air bottles to operate SG PORVs • Manual operation of TDAFW • Spray nozzle (located on site fire truck) for scrubbing fission product release

18. Surry Power StationLong-term Station Blackout With Mitigation Swollen Vessel Water Level Response

19. Surry Power StationShort-term Station Blackout With Mitigation (Emerg. CS) Swollen Vessel Water Level Response

20. Surry Power StationShort-term Station Blackout With Mitigation (Emerg. CS)

21. Surry Power StationISLOCA With Mitigation Swollen Vessel Water Level Response

22. Surry Power StationISLOCA With Mitigation ISLOCA mitigated using Second Unit RWST

23. Mitigative Measures Sensitivity Analysis Without mitigative measures • Long term SBO • Core damage at 16 hrs • Containment failure at 45 hrs (increased containment leakage) • Public evacuation begins at 2.5 hrs • Short term SBO • Core damage at 3 hrs • Containment failure at 25 hrs • Public evacuation begins at 2.5 hrs • ISLOCA • Release scrubbed in flooded Aux building room • Non-mitigated analysis ongoing • SGTR • Unsuccessful mitigation not considered credible • >40 hrs to core damage and offsite release

24. Surry Power StationLong-term Station Blackout Without Mitigation Swollen Vessel Water Level Response

25. Surry Power StationLong-term Station Blackout Without Mitigation

26. Surry Power StationShort-term Station Blackout Without Mitigation

27. Surry Station BlackoutsCompared to SST-1

28. Peach Bottom Long Term Station BlackoutCompared to SST-1

29. Summary • SOARCA study completing evaluation of Surry and Peach Bottom plants • Releases for unmitigated accident vastly reduced and delayed in time compared to SST-1 • Mitigation shown to capable of terminating accidents • Sequoyah analysis getting underway • Uncertainty analysis and peer review planned